Glass vitrification is a preferred method of immobilizing nuclear waste produced by nuclear power plants (spent nuclear fuel), and also legacy defense waste in nuclear countries such as the United States, former Soviet Union countries, France, and United Kingdom. In this vitrification method, the nuclear waste such as spent nuclear fuel and high to low level waste streams, are mixed with borosilicate, aluminum phosphate, or iron phosphate glass fits, and heated to a vitrification temperature that ranges from about 1000 to 1500° C. The resulting melt solidifies into a glassy mass that encapsulates the radioactive and hazardous contaminants present in the waste, and immobilizes them for millions of years. This vitrified mass can be stored in repositories or in special vaults built for that purpose.
The glass vitrification method has drawbacks, however. The nuclear wastes contain very high activity fission products isotopes such as iodine (I-127), cesium (Cs-134 and Cs-137), technetium (Tc-99), and often strontium (Sr-90), which are volatile and become part of the off-gases at their boiling point during vitrification. Technetium often is present in nuclear waste materials as pertechnatate (TcO4−), which is volatile and can evaporate at a temperature as low as about 140° C. Technetium dioxide (TcO2) evaporates at a temperature of about 800 to 900° C. Cesium, strontium, and iodine also go into vapor phases at relatively low temperatures, and thus their radioisotopes cannot be immobilized in the glass matrix and need to be captured as a separate secondary waste stream and then immobilized in a separate procedure.
Chemically bonded phosphate ceramic (CBPC) based stabilization is based on a method for room temperature immobilization of fission products and works very satisfactorily in many cases. However, several issues have hindered practical utilization this process. For example, most nuclear countries recognize only traditional vitrification for treatment of most of the high level waste streams and some low level waste streams. Acceptance of alternative technologies by the stakeholders and other interested parties has been limited, in spite of benefits from such alternatives. Currently, there are no official waste acceptance criteria and accepted testing procedures for alternative waste-forms, such as CBPCs, that are produced at room temperature. In addition, vitrification results in a significant waste volume reduction as the volatile compounds, including water, present in the waste are driven away. In any room temperature stabilization technology the volatile compounds remain in the waste-form, and as a result, there is no volume reduction. In fact, depending on the waste stream or other factors, there can be an increase in the volume. Storage of high volume waste-forms is expensive, and over a period of many years the cost of vitrification and storage may turn out to be cheaper than room temperature stabilization despite initial high cost of vitrification.
The issue of volatile fission products is not limited to legacy defense waste or other liquid wastes only. Nuclear power plants generate spent nuclear fuel, which also faces the same problem during vitrification. In addition, to improve the efficiency of the power plants and also to reduce the volume of the waste stream, several technologies are being pursued to recycle the spent fuel and extract more power out of it. As more power is extracted, more fission reactions occur, thus producing more fission products (including volatile fission products). Table 1 illustrates Cs-137 concentration in spent fuel from a PWR reactor at various burn-up levels based on information in NUREG/CR-6798, “Isotopic Analysis of High-Burnup PWR Spent Fuel Samples From the Takahama-3 Reactor”, prepared by C. E. Sanders, I. C. Gauld, ORNL (2003).
TABLE 1Burn-up level (Gwd/MTU):7.7914.324.3535.4247.25Cs-137 (g/TU):28054093013501760
The burn-up level results in reduction of actinide energy in the spent fuel, but because of more fission reactions, the fission product concentration increases and hence the problem of immobilization of Cs-137 in the waste stream gets worse. For example, raising the burn-up level from 7.79 Gwd/MTU to 47.25 Gwd/MTU increases cesium content from 280 g/MgU to 1760 g/MgU. From the data in the NUREG/CR-6798 referenced above, production of volatile fission product appears to be directly proportional to the burnup level. It is expected that future plants will have even higher burnup levels. Thus, by raising burn-up level, the level of fission products in it will be higher even though total spent fuel will be reduced. Thus, the issue of volatile fission products will gain more importance in the near future.
There is also another safety and security issue with Cs-137. This isotope is used in commercial radioactive sources used for blood sterilization, medical and industrial research, and instrument calibration. Commercial Cs-137 sources are produced by packing radioactive cesium chloride (CsCl) powder in capsules that are then introduced in irradiator equipment. Cs-137 has an extremely high specific activity (87 Ci/g) and is readily dispersible in groundwater. Even a small amount can contaminate a large body of water. The IAEA has identified Cs-137 sources as a global security threat and NRC policy statement calls for development of safe sources by using safer matrices.
The methods and materials described herein address the issues described above by providing a method of immobilizing waste materials, such as radioisotope wastes, in a non-dispersible, and hence safe, vitrified CBPC article that does not require altering the capsule design or hardware utilized in conventional CBPC immobilization. The resulting ceramic has negligible solubility, is difficult to fragment, and is stable even at high temperatures.